Scavenger and process of scavenging



United States Pateiif 2,934,404 r SCAVENGER ANDPROCESS or SCAVENGING CarlM. Olson, Newark, NJ., assignor to the United States of America as represented by the United States Atomic Energy Commission No Drawing. Application November 2, 1945 Serial No. 626,431

I Claim. (Cl. 2314.5)

and the like radioactive contaminants. More particularly this invention concerns a separation and concentration procedure involving the use of a titanium compound as a scavenger whereby improved decontamination may be accomplished.

As described herein, the' isotope of element 94 having amass of 239 is referred to as 94 and is also called plutonium, symbol Pu. In addition, the isotope of element 93 having a mass of 239 is referred to as 93 The term plutonium values used herein is to be understood as denoting the element'either in its free state or in the form of a compound, unless indicated otherwise by the context.

Elements 93 and 94 may be obtained from uranium by various processes which do not form a part of the present invention including irradiation of uranium with neutrons from any suitable neutron source, but preferably the neutrons used are obtained from a chain reaction of neutrons with uranium.

Naturally occurring uranium contains a major portion of U a minor portion of U and small amounts of other substances such as UX and UX When a mass of such uranium is subjected to neutron irradiation, particularly" with neutrons of resonance or thermal energies, U by capture of a neutron becomes U which has a half life of about 23 minutes and by beta decay becomes 93 The 93 has a half life of about 2.3 days and by beta decay becomes 94. Thus, neutron irradiated uranium contains b oth93 and 94 butiby storing such irradiated uranium for a suitable period of time, the 93 is converted almost entirely to 94 ln'addition to the above-mentioned reaction, the reaction of neutrons with fissionable nuclei such as the nucleus of U results in the production of a large number of radioactive fission products. As it is usually undesirable to produce a large concentration of these fission products which must, in view of their high radioactivity,

. be'separated from the 94 and further as the weight of radioactive fission products present in neutron irradiated uranium is proportional to the amount of 93 and 94 formed therein, it is generally preferable to discontinue the irradiation of the'uranium by neutrons when the combined amount of 93 and 94 is equal to a certain predetermined percentage by weight of the uranium mass.

A number of processes have already been proposed for accomplishing decontamination and the separation and concentration of Pu. Certain of these processes are generically known as the bismuth phosphate" process and the wet fluoride process. These processes are the in- JVention of others and the details of the processes are described in copending applications as, for example, application S er..No. 519,714, now Patent No. 2,785,951, to be referred to hereinafter. Consequently all of the details ice of operation of the aforementioned processes are not described herein.

As indicated above, neutron-irradiated uranium contains a substantial amount of fission products which, in view of their radioactivity, are hazardous and present difficulties unless removed. The amount of fission products present can be reduced substantially by the bismuth phosphate process, by other existing processes, and by repetition and combination of these processes.

However, it has been noted that while processes as exemplified by the bismuth phosphate type of treatment when first employed are particularly eifective in eliminating a large amount of-activity, repetition'thereof will not completely remove all of the activity. This presumably may be due to thefact that certain activities are not sensitive to, do not have'an affinity for, or otherwise are not affected by the bismuth phosphate carrier. Hence, repeti tion of cycles using the bismuth phosphate carrier does not entirely eliminate these specific activities. The same condition apparently, also prevails, with other carriers which in their initial use will carry substantial amounts of activity but which will not entirely eliminate the activity. V

It has also been found that there are certain reagents which may be employed in conjunction with by-product carrier precipitate formation for improving decontamination. These reagents not only increase the amount of fission product activity carried down, but apparently also,

function by carrying down certain specific activities which have heretofore been diflicult to remove. The use of such. supplemental additions are referred to as scavenger additions. 1 For example, as described in detail in companion application Serial Number 614,615, filed September 5, 1945, in the name of Thomas H. Davies, it has been found that supplemental additions comprising titanium dioxide and the like improve decontamination by acting as scavengers. The use of 'such reagents as described in said companion application permit the elimination of a greater amount of the. gamma contamination. However, the forms and sources of such reagents as titanium dioxide heretofore used, while carrying down the activity, have given precipitates presenting some difiiculties of separating by centri fuging, washing, and the like.

In accordance with the present invention, I have found that subpigment grade, uncalcined, preformed titanium dioxide functions in an improved manner as a scavenger.- While the preceding description has been directed to the separation of activity in the recovery of Pu, values as this is a common situation in which the invention would be used, it may be used in other instances. For example; it may be desirable to isolate gamma activity for special purposes, or to isolate certain fission componentsfrom materials containing no Pu values and the present invention may be employed in these instances.

The meaning of the terms decontaminated, scavenging, scavengers, product precipitate, by-product precipitate, bismuth phosphate process, extraction and other similar terms will be further apparent as the description proceeds.

The invention has for one object to provide improvements in processes for the separation and recovery of Pu values.

, Another object is to provide a method of treating materials containing Pu values whereby decontamination may be better and more easily accomplished.

. Another object is to provide a process for the separation and recovery of Pu values by a procedure which includes scavenging. Still another object is to provide a procedure in the separation and recovery of Pu values whereby specific activities may be removed.

Still another object isto provide a scavenging method Patented Apr. 26, 19 60 which may be used in conjunction with known processes for extracting and decontaminating materials containing Pu values.

A still further object is to provide a scavenger which lends itself to easier resluIryin-g, washing, and separation operations.

Another object is to provide a scavenging process which may be carried out in existing equipment without change along with known processes or with a minimum of change.

Other objects will appear hereinafter.

I have found that by supplementing by-product decontamination steps by a step comprising employing a subpigment grade, uncalcined titanium material as a scavenger decontamination and isolation of certain fission activity may be obtained in an easier and better defined manner than with reagents heretofore used. In particular I have found that by means of the use of subpigment grade, uncalcined Ti in accordance with the present invention troublesome gamma activity present with the Pu values may to a large extent be eliminated. The TiO scavenging may be coupled with fluosilicic acid complexing in the product precipitation step as will be 1 described in further detail hereinafter.

In order to illustrate the environment in which my invention may be used, certain typical general process steps are now described, together with general designations of points where my invention may be used and other associated information.

The type of irradiated materials such as materials containing Pu values which may be treated are any of the usual mixtures or solutions thereof encountered generally. For example, such materials are described in companion application Ser. No. 519,714, now Patent No. 2,785,951, aforementioned. These materials would, for example, comprise an inorganic acid solution containing the Pu values, the radioactive materials which it is desired to eliminate, and various other constituents. In general, the initial materials would have been subjected to conventional processing comprising extraction and concentration for the elimination of some of the extraneous substances. That is, a standard product precipitation would have been applied under (r) conditions and the carrier containing Pu together with that radioactivity which had also been carried along would have been redissolved and subjected to a conventional by-product precipitationunder (0) conditions for the elimination of radioactivity. The scavenging of the present invention would usually be applied in conjunction with such a by-product precipitation for increasing the decontamination obtainable thereby. 7

In the event it is desired to isolate only a certain activity without reference to Pu recovery, an ether extract, the water exchange therefrom or comparable liquid may be treated with the scavenger of the present invention for carrying down certain activities to be isolated The types of carrier precipitates involved and related subject matter are described in application Ser. No. 519,714, now Patent No. 2,785,951, filed January 26, 1944, Thompson and Seaborg, and reference is made to that application for further disclosure, details thereof being omitted from the present disclosure except where necessary to an understanding of the present invention. As set forth in said application, it has been discovered that the element Pu, under the proper conditions, can exist in aqueous solution in a number of oxidation states and the ability of carrier precipitates to carry plutonium is dependent on the oxidation state of plutonium.

It has been found that plutonium present in a solution acidified with an acid such as nitric, sulphuric, or hydrochloric and in the absence of strong oxidizing agents, or preferably in the presence of a reducing agent such as sulphur dioxide or ne oxalic acid, formic acid, can be substantially carried out of solution by co-precipitation with some insoluble rare eanth compound, such as 2,934,404 a f r lanthanum or cerium fluoride, or with certain insoluble phosphates, such as the phosphates of-bismuth, lanthanum, and zirconium. Rare earth iodates and oxalates are also efiicient materials for carrying plutonium out of solution under these conditions. Thorium fluoride, iodate or oxalate also are eflicient carriers for plutonium under these conditions. The plutonium when it is in this carriable condition, generally referred to as Pu probably is in the oxidation state of +4 and in some instances may be in a +3 state. The oxidation state of plutonium in this condition may be referred to as the fluoride insoluble or phosphate insoluble state.

In sharp contrast to this fluoride insoluble or phosphate insoluble state of plutonium, it has been found that the addition to the acid solution of a strong oxidizing agent, such as peroxydisulphate ion (8 0 plus silver ion (Ag++) as a catalyst, oxidizes plutonium to an oxidation state or states in which it is not carried by rare earths or thorium fluoride, iodate or oxalate or by the phosphates of bismuth, zirconium, or lanthanum, when these are precipitated from the solution. Under these conditions, the plutonium is oxidized to state greater than +4 as the +5 or +6, oxidation states and may be generally referred to as Pu The particular valence state of the plutonium in this condition is not important in this invention, provided it is phosphate and fluoride soluble, and such plutonium in solution may be referred to as being in the fluoride-soluble or phosphate soluble" oxidation state. It has been found that a number of oxidizing agents are capable of oxidizing the plutonium from the fluoride insoluble or phosphate insoluble oxidation state to the fluoride soluble or phosphate soluble" state, and the oxidizing potential required for this change is more negative than ---1.0 to 1.4 volts on the Latimer scale of potentials. For example, it has been found that permanganate, periodate, dichromate and ceric ions in acid solutions can be used to effect this oxidation. It has also been found that reducing agents more positive than the above-mentioned potential range can effect reduction from the fluoride soluble" or phosphate soluble oxidation state to the fluoride insoluble or phosphate insoluble state.

A standard bismuth phosphate by-product precipitation step alone may eliminate one-half or more of the gamma activity dependent on the initial intensity and related factors. On the other hand, a comparable by-product precipitation cycle conducted in the same manner but supplemented by the use of a preformed titanium dioxide of a subpigment grade, as a scavenger and otherwise carried out in accordance with the present invention removes a much greater amount of the gamma activity.

Such titanium dioxide has a granular structure which permits reslurrying and the washing of the scavenger precipitate. This is advantageous in that the titanium dioxide scavengers heretofore used have formed precipitates which were diflicult to separate by centrifuging.

As indicated the special titanium dioxide scavenger described herein may be used in the same manner as scavengers which have been previously employed. That is, a bismuth phosphate or other by-product precipitation step may be supplemented by the addition of a preformed titanium dioxide scavenger which scavenger precipitate may be more easily centrifuged out and is also more susceptible to treatment by washing and slurrying. It is also proposed in accordance with the present invention to couple the use of the aforementioned titanium dioxide scavenger with fiuosilicic acid complexing. That is, when a product precipitation, that is precipitation with a carrier under (1') conditions, is subsequently applied to a scavenged solution, a certain amount of the scavenger, not completely separated in the preceding centrifugation may tend to carry along with the precipitate carrying product. By the addition of a small content of fluosilicic acid to the solution in which the product precipitate is made, these. scavenger fines which might tend to-be carried along are dissolved and remain in solution thereby not contaminating the precipitate containing Pu values.

The titanium compounds preferred in the present invention are most readily obtained by processes of the type described in U.S. Patents Reissue 17,429 and 17,430 to Joseph Blumenfeld or in Reissue 18,790 to Mecklenburg. While these processes are primarily directed toward production of titanium dioxide of pigment grade, they may be modified to give coarser, granular materials by conducting the hydrolyses at lower concentrations than specified therein. As an example, instead of precipitating the titanium compound from a solution containing approximately 200 grams per liter Ti0 as described in Reissue 18,790, page 2, column 2, one can obtain a coarse product suitable for the decontamination process of the present invention by hydrolyzing at 100 grams per liter TiO A still further understanding of my invention may be had by a consideration of the example which follows:

Example I In accordance with this example a conventional acid solution of irradiated U materials, such as of a type as described in Ser. No. 519,714, now Patent No. 2,785,951, was treated by the usual BiPO cycles. At the conclusion of a by-product precipitation cycle, prior to applying the scavenging treatment of the present invention studies were made relative to the gamma activity.

The standard bismuth phosphate by-product precipitation alone brought down approximately 50-55% of the original gamma activity. On the other hand, it was found, in this example, that a comparable by-product precipitation cycle employing a comparable bismuth phosphate carrier supplemented by the use of 2%-5% slurry of preformed, uncalcined titanium dioxide of subpigment grades as a scavenger brought down more than 75% of'the.

gamma activity. Also, this scavenger precipitate was relatively easily separated by centrifuging. The separated scavenger precipitate was washed relatively easily in the centrifuge, three times with water, the washings being combined with the centrifugate liquid, thereby mimmizing the possibility of loss of Pu values on said precipitate. While in this example the scavenger was incorporated supplemental to the by-product precipitate, it will be noted that scavenging of the present invention may be carried out either before, simultaneously, or subsequent to theprimary bismuth phosphate or other by-product precipitation cycle. Although the scavenger of the present invention may not remove a greater amount of contamination than scavengers heretofore used, nevertheless the scavenger exhibits advantages particularly in large scale industrial use. The preformed scavenger of the present invention may be easily incorporated, preferably with agitation with the liquid to be scavenged at whatever temperature the liquid is at. While a fraction of a gram of scavenger per liter of liquid to be treated is sufiicient, excess of the preformed scavenger may be used as it easily separates out on centrifugation. Also, as indicated, the separated scavenger may be readily reslurried due to its granular structure.

As described above, I have found that in general gamma activity and also certain other fission activities may be removed by means of preformed, uncalcined titanium dioxide of subpigment grade as a scavenger. The solutions which are especially suitable for treatment result from dissolving irradiated uranium in nitric acid and applying the usual bismuth phosphate cycle thereto with the scavenger treatment of the present invention applied at about the time of a by-product precipitation under (0) conditions. Or the ether extract or aqueous exchange liquid from such nitric acid solutions of the irradiated material may be treated. However, the exact type of solution treated is not a limitation upon the present invention. It has been found that in instances where plutonium is present, plutonium recovery is as efficient where the scavengers of the present invention are used as where a carrier, such as bismuth phosphate, is used alone. Usually a fraction of a gram of scavenger per liter of liquid treated 'is sut'ficient. However, up to 5 grams per liter may be used. By the term scavenger I intend to embrace in particular the preformed, uncalcined TiO described above and also any of the various compounds such as alkali or alkaline earth titanates which remain substantially insoluble in the phosphate-containing solution and which likewise function to remove activity either in the presence or absence of Pu values as described. The exact details respecting forming BiPO, or other carrier precipitates, oxidizing or reducing solutions of irradiated materials and similar details are not a limitation on the present invention and are the invention of others. As already indicated, my scavengers and scavenging treatment may be employed in conjunction with various carriers of which BiPO has merely been described as a typical example.

It is to be understood that all matters contained in the above description and examples are illustrative only and do not limit the scope of this invention as it is intended to claiirrthe invention as broadly as possible in view of the prior art.

I claim: 7

A process for separating plutonium values from an acidicaqueous solution containing said values together with fission products which comprises forming a bismuth phosphate precipitate in said solution while maintaining said plutonium in an oxidation state greater than four, whereby a portion of said fission products are separated from said plutonium, contacting the resulting partially purified solution with preformed, uncalcined, granular.

titanium dioxide, whereby additional fission products are separated from said plutonium, removing said bismuth phosphate precipitate and said titanium dioxide from the resulting mixture, and contacting the resulting purified plutonium-containing solution with fluosilicic acid.

References Cited in the tile of this patent UNITED STATES PATENTS 2,785,951 Thompson et al Mar. 19, 1957 

